正常运行工况熔盐堆主回路衰变热特性研究

基于Mathematica 7.0建立了熔盐堆(Molten Salt Reactor,MSR)主回路系统衰变热流动模型,并与参考程序ORIGENS在静态燃耗下的计算结果以及熔盐实验堆(Molten Salt Reactor Experiment,MSRE)衰变热结果进行了初步验证,相对偏差分别在±4%和±2.76%的范围内符合较好。对2 MW液态燃料钍基熔盐实验堆(Thorium Molten Salt Reactor-Liquid Fuel 1,TMSR-LF1)正常运行工况下主回路系统管道及设备内的衰变热分布进行了定量分析。结果表明:启堆达到满功率和设定流量后约20 s各区域衰变热快速积累,随后便开始平缓上升并趋平衡。平衡时堆芯活性区衰变热占总衰变热的46.7%,上腔室、热管段#1、主泵、热管段#2、换热器、冷管段及下腔室区域分别占比31.8%、1.21%、14.6%、0.89%、2.21%、1.67%和0.94%。所建立的分析方法及结论可为熔盐堆主回路系统的热工水力安全分析、余热排出系统设计、反应堆功率调节与安全控制提供重要参考。     

熔盐堆高温熔盐泵液下轴承表面织构分析与优化

高温熔盐泵是使用熔盐冷却剂热工回路的"心脏",为回路中的熔盐强迫循环提供动力。目前熔盐堆使用的高温熔盐泵采用悬臂式结构,其较长的液下主轴需要采用滑动轴承支撑。在轴承表面加工出具有一定尺寸和规则排列的几何形状,是改善轴承表面摩擦学特性提高承载能力一种有效手段。本文采用有限体积法对高温熔盐泵轴承进行研究,以光滑轴承作为对比项,分析了长条织构、圆形织构、正方形织构和十字形织构对轴承承载能力的影响,通过优化织构的形状、密度、深度等参数,得到长度8 mm、宽度3 mm、深度20μm的十字形织构,这种织构形式的轴承承压能力最好,可以为高温熔盐泵轴承设计提供有效的理论指导。     

TMSR熔盐回路的漏热功率在线监测软件

简述了钍基熔盐堆核能系统(TMSR)熔盐回路漏热功率在线监测软件功能和开发过程。基于EPICS构架,通过采用分段式的实时温度数据采集与在线处理的方法,首次实现了TMSR信号实时处理与监测功能,完成了用于实现TMSR熔盐回路漏热功率在线监测软件的开发,为TMSR监控系统的后续功能开发和TMSR正常运行提供了可靠的技术支持和保障。     

基于堆芯线性化模型的液态熔盐堆功率控制研究

液态熔盐堆中熔盐燃料依托主泵驱动在一回路中流动,在流动过程中造成了反应性损失,直接引起堆芯功率变化。考虑到熔盐燃料流动对堆芯功率控制的影响,建立了堆芯非线性模型,并对模型进行线性化处理。基于堆芯线性化模型,采用线性二次型高斯/回路传输（LQG/LTR）技术设计堆芯功率控制系统。以熔盐实验堆为例,开展堆芯反应性扰动控制研究。结果表明,采用堆芯线性化模型和LQG/LTR技术可以实现对液态熔盐堆堆芯功率的控制。      

小型模块化熔盐堆误提棒ATWS事故分析

误提棒未能紧急停堆（Anticipated Transient Without Scram,ATWS）事故是熔盐堆的超设计基准事故之一，以125 MW液态熔盐堆为研究对象，采用RELAP5-TMSR(Reactor Excursion and Leak Analysis ProgramThorium Salt Reactor)程序，针对误提棒ATWS事故，选取三种停堆策略分析反应堆功率和熔盐温度等关键参数的变化。此外对反应性引入价值、提棒速度和温度系数等若干重要因素也开展了相应的敏感性分析。分析结果表明：维持一回路主泵运行、关闭二回路主泵和三回路风机的停堆策略是三种策略中堆芯熔盐温度最低的；在仅维持一回路主泵运行的情况下，温度极值与反应性引入价值、引入速率及温度反应性系数密切相关，温度峰值随反应性引入价值和提棒速度的增加而增大。     

无铍钍基熔盐堆堆芯设计与安全研究

为解决传统熔盐堆在核燃料增殖、安全性等方面的不足,提出了采用氧化铍慢化剂、无铍(BeF2)燃料熔盐的新型钍基熔盐堆(TMSR)堆芯设计。在此基础上,利用多物理计算程序开展了TMSR稳态及瞬态初步安全特性分析。通过对反应堆启动、熔盐泵超速及降速、丧失热阱等典型瞬态的计算,分析了各种工况下堆芯功率与温度的变化情况。结果表明,在各种运行瞬态及事故情况下,新型的TMSR设计具有良好的安全特性。     

先进压水堆核电厂安全壳内滤网设备压损研究

在AP1000机组中,安全壳内滤网设备用于过滤失水事故后循环冷却水中的杂质,确保非能动堆芯冷却系统(PXS)正常运行。压损是滤网设备重要性能指标,滤网设备压损不能超过规定限值,以保证事故后有足够的循环冷却水可以冷却堆芯。滤网设备压损分为过滤部分压损和流道部分压损,文中通过试验手段和模拟计算的方法分别得出了一种新型滤网设备的各部分压损值。研究结果表明,在电厂极限工况条件下,该新型滤网设备过滤部分最大压损为0.329kPa,流道部分最大压损为0.636kPa,总压损为0.965kPa,远远小于1.72kPa的设计要求值。本研究结果为此种新型滤网设备下一步投入实际使用奠定了坚实基础。     

液态熔盐堆运行安全特性初步研究

液态燃料反应堆与固态燃料反应堆相比,原理上有较大不同。液态熔盐堆中由于燃料流动带走缓发中子先驱核在堆外衰变导致堆芯反应性降低,且裂变产物在堆外回路中衰变也会引起一回路发热。本文使用熔盐堆中子动力学程序Cinsf1D探讨2 MW熔盐堆的临界动力学特性和安全特性,研究零功率临界下不同熔盐流速启泵和停泵导致的缓发中子先驱核流失所需改变的控制棒棒位。同时还计算了2 MW恒定功率情况下稳态运行及降低流速时一回路温度分布,并模拟了2 MW额定功率下停泵事件。停泵后由于缓发中子损失减少反应堆功率先缓慢增加,然后迅速降低到接近余热水平。停泵后堆芯温度缓慢增加后稳定在安全值以内,说明熔盐堆具有本征安全性。     

液态燃料熔盐堆三维动力学程序开发及验证

液态燃料熔盐堆的燃料熔盐在一回路中循环流动,一回路高温熔盐既是燃料,又是冷却剂,大部分核裂变能直接释放在燃料熔盐之中。随着燃料熔盐流动,一部分缓发中子先驱核（Delayed Neutron Precursors,DNP）在堆芯外一回路中衰变引起反应性损失。液态燃料熔盐堆中子物理与热工流体紧密耦合,传统固态燃料反应堆堆芯核热耦合程序不再适用于液态燃料熔盐堆。针对液态燃料熔盐堆特点,建立了包含带对流项的DNP输运方程和带热内热源热工流体方程的液态燃料熔盐堆动力学模型,并基于节块展开法,开发了堆芯三维动力学程序ThorCORE3D。使用美国橡树岭国家实验室建造运行的熔盐实验堆（Molten Salt Reactor Experiment,MSRE）稳态和瞬态实验基准题,对ThorCORE3D程序进行了初步验证。结果表明：ThorCORE3D程序计算值与MSRE实验值吻合良好,适用于液态燃料熔盐堆稳态设计与瞬态分析。     

一种热管熔盐堆塔式温差发电系统设计及分析

熔盐堆作为第四代先进反应堆的重要堆型之一，以高沸点熔盐为核燃料熔融载体，具有高温输出、常压操作等特点。而基于温差发电的热管熔盐堆，兼具了熔盐堆、热管和温差发电的优势，具有输出温度高、热电转换效率高、结构简单及安全可靠等优点，在能源系统领域具有极大的优势，是外太空及深海探测任务的理想能源。但因堆芯熔盐低热导率而形成的热管密集排布给热管冷凝段的温差发电传热设计带来了难题。针对该堆型设计需求，本文提出适于熔盐堆的热管-温差发电耦合系统结构并进行了传热分析。堆芯热管冷凝段采用塔式温差发电系统结构设计，整体热端座与堆芯热管冷凝端相配合，形成从下至上的第1层至第N层热段套；冷端座套置于热端座外，内设冷端热管通道；热端座的外侧壁与冷端座的内侧壁之间贴有温差发电片，发电片间隙采用保温棉减少漏热。采用Ansys Workbench开展了适于热管熔盐堆的4层塔式温差发电系统传热仿真模拟，分析表明：系统运行的高温热管最高温度为696℃时，整体塔座温度分布均匀，热量有效利用率大于96%，系统漏热量小于4%，发电片两侧温差大于490℃，利于提高热电转换效率，设计具有可行性，有利于推动温差发电在热管熔盐堆中的应用...     

高温熔盐试验回路系统设计及验证研究

The Thorium Molten Salt Reactor (TMSR) project plans to construct a 2 MWt liquid fuel molten salt reactor. After successful R&D of the proto-type equipment like pump, heat exchanger and freeze valve, construction of a high temperature fluoride loop is designed and constructed to test them. The rated power of heat exchange is designed to be 200 kW, and the flow rate is set as 15 m3/h by the caloric equation. Hence the pipe caliber chose to be DN50. The numerical computation by Fluent revealed that the pressure drop of the loop system is about 155 kPa, and the overdesign of the pump head is set as 20 m accordingly. To avoid thermal stress concentration of the pipeline, both the electric heater and radiator are designed to be fixed on a flexible universal sphere support, except the pump which is anchored on steel support directly. The structural design of the equipment and the loop hase been verified by several tests with an accumulated operating time near 4000 hours. The actual pressure drop of the loop is about 110~120 kPa, which still needs to be accurately measured by differential manometer in future. The analysis of the impurity of the molten salt shows that the contents of Cr and Mo increased by two orders of magnitude after loop operation, which means that an obvious material corrosion occurs. The level of moisture and oxygen, which is the main reason that cause fluoride corrosion, should be controlled considerably lower than the original design level of 100 μL/L.     

The protective performance of a molten salt frozen wall in the process of fluoride volatility of uranium

The fluoride volatility method(FVM) is a technique tailored to separate uranium from fuel salt of molten salt reactors. A key challenge in RD of the FVM is corrosion due to the presence of molten salt and corrosive gases at high temperature. In this work, a frozen-wall technique was proposed to produce a physical barrier between construction materials and corrosive reactants.The protective performance of the frozen wall against molten salt was assessed using FLiNaK molten salt with introduced fluorine gas, which was regarded as a simulation of the FVM process. SS304, SS316 L, Inconel 600 and graphite were chosen as the test samples. The extent of corrosion was characterized by an analysis of weight loss and scanning electron microscope studies. All four test samples suffered severe corrosion in the molten salt phase with the corrosion resistance as: Inconel 600 SS316 L graphite SS304. The presence of the frozen wall could protect materials against corrosion by molten salt and corrosive gases, and compared with materials exposed to molten salt, the corrosion rates of materials protected by the frozen wall were decreased by at least one order of magnitude.     

熔盐自然循环回路热损失功率实验及计算

熔盐自然循环回路是为研究熔盐的自然循环特性,支持先进熔盐堆非能动安全系统设计而建造的实验装置。熔盐在回路中的热量损失对于自然循环的建立和保持具有重要的影响。本文以熔盐储罐为代表部件,通过实验得到了其热量损失的数据,并利用数值模拟的方法,计算了不同温度下储罐各部分的热损失,分析了储罐热损失规律,拟合得到了热损失随温度及时间的变化关系。对比熔盐在不同温度下热损失的实验值和计算值,发现两者吻合良好,相对误差均小于10%。分析结果表明,内插式电加热器是储罐主要热损失途径之一,并导致了熔盐的温度分层。     

熔盐冷冻壁应用中关键工艺影响因素研究

在熔盐堆燃料干法处理流程中,处理设备面临着严重的材质腐蚀问题。熔盐冷冻壁技术被视为保护相关设备耐受化学腐蚀的有效方法,而冷冻壁厚度的稳定控制是干法处理流程应用冷冻壁技术实现处理工艺目的的关键。基于自行研制的冷冻壁实验装置,模拟了干法处理中熔盐冷冻壁的应用工况,考察了导热油进口温度、熔盐初始温度、加热器功率、冷冻壁初始厚度对冷冻壁厚度变化的影响,得到了各个因素的影响规律,并总结了最佳的应用工艺条件。利用热流量的变化分析了冷冻壁厚度变化的原因:热流量越大,冷冻壁厚度减小量越大,达到平衡时,热流量越大,冷冻壁平衡厚度越小。通过实验数据拟合得到了线热流密度与冷冻壁平衡厚度的关系式,平均相对误差11.2%。     

Corrosion behavior of Hastelloy-N alloys in molten salt fluoride in Ar gas or in air

The effects of air on the corrosion of Hastelloy-N alloys in molten salt coolant containing fission product elements were investigated to determine the safety of structural materials in high-temperature reactors cooled with fluoride salt. Corrosion tests of Hastelloy-N in the molten fluoride salt FLiNaK in an alumina crucible and a graphite crucible under argon gas or air were performed at 773–923 K for 100 h. The depth of corrosive attack, as well as the extent of chromium and molybdenum depletion, increased with increasing temperature. The extent of Hastelloy-N corrosion in molten salt under air was significantly greater than under argon gas. The effect of adding the impurity cesium iodide to molten salt containing nuclear waste fuel on the corrosion behavior was negligible.     

一体化小型氟盐冷却高温堆瞬态特性分析

根据下一代核能系统的发展目标,提出了采用自然循环的一体化小型氟盐冷却高温堆的概念。利用修改后的RELAR5-MS系统分析程序,建立了一体化小型氟盐冷却高温堆模型,并得到其稳态特性参数。在此基础上,对其在满功率运行状态下的反应性引入事故和失热阱事故进行了分析。分析计算表明,在反应性事故工况下,由于自然循环的存在,堆芯冷却剂流量随着堆芯温度发生动态变化,最终达到新的稳态,燃料棒和冷却剂温度均处于安全限值范围内。在失热阱事故下,反应堆负反馈的特性使得堆芯功率逐渐降低并实现自动停堆,即使不考虑余热排出系统的作用,燃料组件和冷却剂温度上升缓慢,在140 h内,燃料棒和冷却剂温度均处于全限值范围内。结果表明,一回路采用自然循环冷却的一体化小型氟盐冷却高温堆具有良好的固有安全性。     

高温原位氟化熔盐红外吸收光谱装置研制

钍基熔盐堆是第四代核能系统的候选堆型之一,熔盐由于优异的传热性能与中子特性,被作为冷却剂和燃料盐载体在堆内运行,其微观结构对其物理化学性质有重大影响,因此,研究熔盐结构对熔盐制备净化、腐蚀控制以及熔盐堆的设计、建造和安全运行都有重要的指导作用。红外吸收光谱(Infrared Absorption Spectroscopy,IR)是研究熔盐结构的有力工具之一,但标准仪器无法实现高温(500oC)熔盐的测量。本研究在解决了样品加热、环境气氛控制以及腐蚀等难题后,研制了一套适用于高温氟化熔盐实验的高温原位红外吸收光谱装置。该装置中,加热炉为左右两开式,可以直接放进标准红外光谱仪样品仓内;样品池为组装式,主体是能够耐氟化熔盐腐蚀的镍基哈氏合金,窗片为单晶Si C或金刚石;样品池整体呈倒"T"型,上端密封盖带有进气口和出气口,可实现抽真空或通惰性气体的操作。该装置可以实现25-600oC温度范围测量,波段范围覆盖近红外和中红外(14 700-400 cm~(-1))。通过使用常温水和高温亚硝酸钠的红外光谱实验对该装置的可靠性进行了验证。利用该装置,我们成功地获得了高温氟锂铍(FLi Be)熔盐的红外吸收光谱。     

石墨慢化通道式熔盐堆的稳态热工水力计算模型

针对石墨慢化通道式熔盐堆的堆芯结构,基于COMSOL Multiphysics程序和MATLAB程序建立了堆芯稳态热工水力学计算模型。该模型对堆芯内固体区域的温度分布采用三维热传导方程进行模拟,对通道内熔盐温度采用一维单相流体模型进行计算。固体区域与熔盐通过熔盐通道壁面的对流换热边界建立热耦合。该模型基于平行通道压力损失相等的原则,分配堆芯内各熔盐通道的流量。通过对比RELAP5程序的计算结果,验证了模型对温度和流量分配计算的正确性。针对2 MWt 液态燃料熔盐堆的一种概念设计,分析了堆芯内三维温度分布和通道间流量分配。该模型可精确计算通道式熔盐堆堆芯内稳态温度分布和流量分配,对堆芯的热工水力学设计具有重要意义。     

Activation product transport in a FLiBe-vanadium alloy-HT9 system

An assessment is made of the gamma radiation hazards likely to be found around a fusion reactor heat transfer and tritium breeding loop which employs a vanadium alloy for the blanket and first wall structure and the ferritic-steel HT9 for the remainder of the loop. The coolant/tritium breeding fluid is the molten metallic salt FLiBe. Since the radiation levels near the primary loop components are found to be less than 100 mR/h 3–5 days after shutdown after three years of continuous full power operation, limited hands-on maintenance could be allowed. The very short half-lives of the predominant corrosion products make this result possible and make such a system very attractive.     

Analysis on the neutron kinetics for a molten salt reactor

The neutron kinetics of the molten salt reactor is significantly influenced by the fuel salt flow, which leads to the analysis methods for the conventional reactors using solid fuels not being applicable for the molten salt reactors. In this study, a neutron kinetic model considering the fuel salt flow is established based on the neutron diffusion theory, which consists of two-group neutron diffusion equations for the fast and thermal neutron fluxes and six-group balance equations for delayed neutron precursors. The temperature feedback on the neutron kinetics is considered by introducing a heat transfer model in the core, in which the group constants which are dependent on the temperature are calculated by the code DRAGON. The mathematical equations are discretized and numerically calculated by developing a code, in which the fully implicit scheme is adopted for the time-dependent terms, and the power law scheme is for the convection–diffusion terms. The neutron kinetics is conducted during three transient conditions including the rods drop transient, the pump coastdown transient and the inlet temperature drop transient. The relative power changes and the distributions of the temperature, neutron fluxes and delayed neutron precursors under these three different transient conditions are obtained in the study. The results provide some valuable information for the research and design of this new generation reactor.